Research

• Projects Completed and In-Progress
• Testing Capabilities and Facilities
• Computational Codes
• Recent Publications
• ISNPS Reference Library (Publications 1975-2008)

Projects Completed and In-Progress

ISNPS possesses a wide range of technical expertise, including: design, thermal-hydraulics and neutronics analysis of gas-cooled, liquid metal-cooled and heat pipe-cooled space nuclear reactors; design optimization and performance of heat pipe radiators; thermal management of Space Nuclear Reactor Power Systems (SNRPSs); transient modeling of heat pipes, including the startup from a frozen state; transient operation, safety and autonomous control of fully-integrated SNRPSs; modeling, design optimization and vacuum testing of high-temperature energy conversion devices, such as thermionic diodes, segmented and non-segmented thermoelectric devices, and Alkali-Metal Thermal-to-Electric Converters (AMTECs); and design, optimization, and thermal and stress analyses of segmented and cascaded thermoelectric converters for SNRPSs and Advanced Radioisotope Power Systems (ARPSs). The results of the research conducted at ISNPS since 1984 have been widely published in refereed technical Journals and Proceedings of technical conferences.

Some of the relevant technology and research projects conducted at ISNPS include:

• Design of gas-cooled, liquid metal-cooled and heat pipe-cooled space nuclear reactors; this effort at ISNPS has led to the development of three innovative reactor concepts: the gas-cooled pellet bed reactor (PeBR), the bimodal PeBR for nuclear electrical power and thermal propulsion applications, and the recent liquid metal-cooled, Sectored Compact Reactor (SCoRe) for the avoidance of single-point failure in the reactor cooling system.

• Neutronics design and analysis of SNRPSs; current efforts are investigating the use of Spectral Shift Absorber (SSA) materials, as a passive and effective means to ensure sub-criticality of fast-spectrum space reactors in the event of water/wet-sand submersion with or without core flooding subsequent to a launch abort accident.

• Nuclear fuel design, performance, and chemical and mechanical analyses.

• Thermal management of spacecraft, power systems, and thermal energy storage systems (e.g. employing the melting and freezing of LiF or other energy storage materials, with an understanding of the thermal and change-of-phase processes in microgravity).

• Interaction of lasers with spacecraft structure , and application of Monte Carlo uncertainty analysis for performance assessment and design of future space systems.

• State-of-the-art modeling and design optimization capabilities for heat pipe radiators which use alkali metal (cesium, rubidium, potassium, sodium and lithium) and low temperature (water) heat pipes, including the startup from a frozen state. These capabilities have been extensively verified using test data.

• Design, experimental development and testing of low-temperature heat pipes , such as water heat pipes.

• Modeling of transient and steady-state operations and safety of fully-integrated space nuclear power systems (e.g. the ISNPS models of the SP-100 space nuclear power systems for 100-1000 kWe power levels (SNPSAM) and of TOPAZ-II and other single-cell TFE type space nuclear power systems (TITAM), and the latest Dynamic Simulation Model (DynMo) for SNRPSs developed at ISNPS using the Simulink® platform integrated with Matlab®).

• Autonomous control of space power systems , with application to the SP-100 space nuclear power systems.

• Modeling, design optimization and testing of static energy conversion devices technology for space applications , including Thermionics (TI), segmented and cascaded ThermoElectrics (TE), and AMTEC converters.

• Experimental investigation and performance evaluation of Pluto-Express (PX) vapor-anode, multi-tube AMTEC cells , in collaboration with the AFRL in Albuquerque, Nichols Research, and AMPS. This effort resulted in major improvements in the design of PX-series AMTEC cells that were being tested jointly at AFRL. This testing, modeling and evaluation effort also included the development of thermal conductivity data for Min-K insulation materials and a comparison of its radial and lateral conductance with those of multi-foil insulation for PX cells.

• 3-D transient and steady-state thermal and mechanical analyses of segmented and Skutterudite TE converters to assess the effect of adding different sublimation suppression coatings on the TE converters performance.

• Design and analysis of ARPSs with Cascaded and Segmented TE converters , with complete design and detailed analysis of the TE arrays and their thermal coupling to General Purpose Heat Source Bricks, and the electrical connections needed to achieve the desired load DC voltage.

• Submersion cooling experiments of high-performance computer chips using pool boiling of dielectric liquids such as HFE-7100 and FC-72 from micro-porous surfaces.

Projects Completed

I. Nuclear Fuel Technology
a. Design and analysis of nuclear fuel storage facilities
b. Thermophysical properties and swelling behavior of nitride and carbide fuels
c. Coated particle fuel compact (CPFC) for use in radioisotope heater units (RHUs) and general purpose heat sources (GPHSs)
  
d. Compatibility of hydrogen with carbide and structural materials at high temperatures

II. Nuclear Reactor Safety
a. Analysis of core meltdown accidents and molten fuel-coolant interaction phenomena
b. Nuclear fuel swelling and irradiation behavior during normal operation and severe accidents
c. Debris bed coolability with application to post-accident heat removal

III. Nuclear Reactor Thermal-Hydraulics
a. Flow visualization and modeling of transition from mixed convection to buoyancy-induced turbulence in vertical annuli
b. Pool boiling and critical heat flux from underside of flat and curved surfaces with application to reactor vessel cooling in severe accidents
  
c. Critical heat flux at low flow and low pressure
d. Natural and mixed convection of water in multi-rod bundles and vertical annuli

IV. Space Nuclear Power and Propulsion Systems
a. Design, optimization, modeling, and analysis of space power systems
     i. SP-100
  
     ii. TOPAZ-II
  
     iii. Pellet Bed Reactor (PeBR) for thermal and electric propulsion missions
  
     iv. Bimodal systems for electric power and thermal propulsion
  
     v. Heat Pipes - Segmented Thermoelectric Module Converters (HP-STMCs) space reactor power system
                vi. Scalable AMTEC Integrated Reactor Space Power System (SAIRS)
b. Design and analysis of SP-100 thermoelectric space power system
c. Studies of void formation during the freezing of alkali metal coolants
d. Design and analysis of TOPAZ-II thermionic space power systems
e. Design and analysis of radiation shield and nuclear reactors for space missions and surface power
f. Reactor design with high burnup capabilities

g. Design of heat pipes-cooled fast spectrum reactors
 
  h. Design of heat pipes radiator for space reactor power systems
           i. Materials studies for use in space nuclear power systems
  
     i. Refractory metals
     ii. Mechanically Alloyed-Oxide Dispersion Strengthened Steels
  

V. Planetary Surface Power Systems - Manned Rover and Surface Outpost

VI. Advanced Radioisotope Power Systems (ARPSs)
a. Coated particle fuel compact (CPFC) for use in radioisotope heater units (RHUs) and general purpose heat sources (GPHSs)
  
b. Vapor anode, multitube AMTEC cells for ARPSs
  
c. SiGe/Skutterudites cascaded thermoelectrics for ARPSs
d. Energy conversion options for ARPSs

VII. Terrestrial Nuclear Power
a. Nuclear power plant concepts with dynamic and static combined energy conversion

VIII. Alkali-Metal Thermal-to-Electric Conversion (AMTEC)
a. Design, analysis and testing of multi-tube AMTEC cells for space and terrestrial energy applications
  
b. Design and performance of high-power AMTEC units for space reactor power systems
  

IX. Thermoelectric Converters
a. High temperature Skutterudites thermoelectrics
b. Segmented thermoelectrics
c. SiGe/Skutterudites cascaded thermoelectrics for ARPSs
d. Performance tests of Skutterudite and Segmented thermoelectric converters

X. Thermionic Converters
a. Development of grooved thermionic electrodes for high power density
  
b . Experimental testing and modeling of the operation of low pressure Cs-Ba tacitron with application to temperature and nuclear radiation power conditioning
  

XI. Heat Pipe Technology and Thermosyphons
a. Experimental and modeling of the transient operation of heat pipes including the start-up from a frozen state
  
b. Design, performance and operation limits of heat pipes, including the start-up from frozen state
c. Design and testing of low temperature heat pipes
  
d. Design, performance and operation limits of thermosyphons
     

XII. Boiling Heat Transfer
a. Boiling heat transfer experiments and correlations

b. Microlayer evaporation and its effect on nucleate boiling heat transfer
c. Flow visualization and high-speed bubble motion studies near critical heat flux
d. Effect of material properties and curvature on pool boiling from downward-facing curved surfaces in saturated and subcooled water
  
e. Boiling of dielectric fluids for electronic microchips cooling

XIII. Immersion Cooling of High-Power Electronics

XIV. Convective Heat Transfer
a. Convection heat transfer experiments and correlations
b. Natural convection and application to electronics cooling
  
c. Marangoni convection and phase-change in microgravity
d. Heat transfer in porous media and thermal energy storage modules
e. Application of laser interferometry to flow visualization and heat transfer in natural convection of air
f. Enhanced cooling using impinging and swirl air jets
g. Natural convection of air at high temperatures

XV. Radioactive and Mixed Waste Treatment
a. Interaction of colloids with gas/liquid interface
  
b. Chemical reversibility of chemical reaction between colloids and contaminants
c. Effect of pH and ionic strength on colloids stability
d. Plasma hearth process and application to the treatment of mixed, chemical and radioactive waste
e. Irreversibility of reactions between colloids and actinides at different cations concentrations
f. Effect of subsurface colloids on hydraulic characteristics of porous media
g. Applicability of filtration theory and correlations to shallow subsurface transport of colloids
h. Attachment kinetics of colloids to gas bubbles in subsurface environment
i. RF plasma for decontamination of transuranic elements
  

XVI. Dynamic Energy Conversion

XVII. Flow in Microchannels and Microtubes/Microfluidics

XVIII. Shielding Solar Energetic Particles
a)Interactions of High Energy Solar Protons with Potential Shielding Materials

b)Displacement Damage Dose in Silicon with Different Shielding Materials

XIX. Very High Temperature Reactor (VHTR) Plants

XX. Neutronics

Back to Top

Testing Facilities and Equipment

Experimental research being performed in the Thermal-hydraulics and Energy Conversion Laboratory (TECL) at the University of New Mexico’s Institute for Space and Nuclear Power Studies includes investigations of pool boiling and two-phase flow systems, thermosyphons and heat pipes, enhanced cooling of electronics, and performance tests of ThermoElectric (TE) unicouples, some of which continued for ~ 3700 hours. The TECL laboratory is equipped with a 200 kW DC power supply, two 1.0 kW DC power supplies, one 500 W DC power supply, three fully equipped data acquisition systems, each connected to a PC equipped with the most recent data analysis software (LabView), two chillers for a wide range of temperature control from –25 oC to 150 oC, and two vacuum test facilities.

Vacuum Test Facilities and Denki v2.0 Software The vacuum test facilities are designed for testing high-temperature devices up to 1300 K under high vacuum (~ 10 -8 Torr), or in inert gas such as argon or helium at < 1 atm. These facilities are fully instrumented for either manual or automatic, real-time data collection of voltage, current and temperature measurements during testing. Thermoelectric performance tests have been performed in both facilities in high vacuum and in argon gas environment at pressures between 0.6 and 0.75 atm. Denki v2.0, a control and data acquisition software developed at UNM-ISNPS using the LABVIEW® platform, is capable of controlling heating, cooling and operation conditions of four different test devices simultaneously. The next version under development will be capable of simultaneously controlling up to 8 different devices under test in the four-bell jar vacuum facility. The software allows automatic control of the experiment to ensure preset operation and that safety margins are not exceeded.

Denki can control tests in various ways. The heater power profile can be specified as a function of time, or a thermocouple reading can be set at specified temperature values. For example, the electrical power to the heater can be automatically adjusted to keep the hot side temperature of thermoelectric devices under test constant, while varying the load current during a current-voltage (I-V) sweep. The control software collects and saves data automatically during performance tests at specified intervals (seconds, hours or days). The cold junction temperature of the TE devices being tested is controlled using a refrigerating / heating bath circulator, with temperature control in the range of 25 oC to 150 oC. Higher cold side temperatures up to 400–500 oC are also possible.

The first facility shown below is a general-purpose vacuum test stand designed to test large-scale assemblies or devices. This facility has a large glass bell jar, 18 inches in diameter and 30 inches high, and its own ion pump (230 l/s capacity down to 10 -8 Torr). This facility is equipped with a HP3497 multi-purpose data acquisition/ control unit for voltage and temperature measurements, a 4-channel Tektronic TDS420 digital oscilloscope for very fast transient measurements (~150 MHz), a CAMAC digital interface system, and other auxiliary digital multimeters. The data acquisition/ control unit, the digital oscilloscope and the CAMAC system are connected and controlled by a computer via GPIB (IEEE-488).

Our TECL engineers and researchers have accumulated a great deal of experience inperforming vacuum tests of TE and Thermionic (TI) converters at elevated temperatures up to 1800 K.

The second vacuum facility shown on this page is a four-bell jar system, for testing a number of devices simultaneously, but at four different conditions. The four glass bell jars are 10 inches in diameter and 12 inches high. This facility has its own vacuum ion pump (500 l/s capacity down to 10 -8 Torr). Each bell jar has its own angle valve to isolate it from the rest of the system, for operating at different conditions (hard vacuum or argon, environment, for example) for intermittent examination of test articles, or replacement with new articles while continuing the tests in the other three bell jars. The articles in each jar can be independently controlled using the existing HP3852A multi-purpose data acquisition / control unit for voltage and temperature measurements. The fully assembled four-bell jar vacuum test facility is shown below.

Experimental Pool Boiling Facility

The experimental pool boiling facility at the Institute for Space and Nuclear Power Studies is dedicated to investigating immersion cooling with dielectric liquids on porous graphite and finned surfaces with the application toward electronics cooling applications.  Systematic investigations are carried out using different boiling flat and finned surfaces (copper and porous graphite), at different liquid subcoolings (0 K ≤ Δ T_sub ≤ 30 K), and at different surface orientations (0° ≤ q ≤ 180°), using both FC-72 (C₆F₁₄) and HFE-7100 (C₄F₉OCH₃) dielectric liquids.

A computerized system controls the experiments and records the boiling curves up to the critical heat flux.  A high quality digital camera and a high quality digital video camera record the boiling at the surface in experiments as they run.  This boiling visualization greatly aid the study of the hydrodynamics of the boiling process from the various surfaces used. 

The pool boiling facility has two separate test stations capable of running experiments independently.  Each station is fully equipped with its own test vessel, chiller, power supply, and computer controlled data acquisition system.  The boiling surface and heater are mounted onto a Teflon block to ensure that all the power, generated by a 200 W power supply (Agilent E3634A), is dissipated through the boiling surface.  The chiller (Thermo NESLAB RTE-7) is capable or removing up to 500 Watts of heat and together with the external water bath help to maintain the liquid bulk temperature to within ± 0.5K of the desired temperature.  A control program developed using the LABView software runs the experiments as well as updates the boiling curve during the experiments and records the data onto the computer’s hard drive for further analysis.

___________________________________

Click Here to
View Our Saturation Boiling Video Library
___________________________________

Back to Top

Computational Codes

• Reactor Design and Advanced Fuel Cycle
  
  
  • MCNP5: Monte Carlo N-Particle version 5
    

    MCNP5 is a general Monte Carlo radiation transport code capable of transporting neutrons, photons, and electrons through virtually any materialin various geometries. A versatile program, MCNP is being used in modeling nuclear criticality and the design of space and terrestrial nuclear reactors and radiation shielding applications. ISNPS uses MCNP to analyze the criticality of its space reactor designs to determine that they meet operational and safety criticality requirements. It is also used for modeling shielding for neutrons and photons produced by reactors and radioactive sources.

  • MCNPX: Monte Carlo N-Particle eXtended
    

    MCNPX is a three dimensional Monte Carlo code (computational tools) capable of treating charge particle transport using either nuclear cross section data or nuclear model calculation.  Current version is capable of treating many light particle such as: neutrons, anti-neutrons, protons, anti-protons, photons, electrons, positrons, muons, anti-muons, electron neutrinos, anti-electron neutrinos, positive pions, negative pions, neutral pions, short K-0s, long K-0s, positive kaons, negative kaons, deuterons, tritons, helium-3s, and alphas.  Therefore, MCNPX makES it possible to compute the radiation effects resulting from secondary particles produced by primary particles interaction with shielding materials.  ISNPS is currently using MCNPX to simulate solar proton radiation and computed the particle fluxes (both for the primary and secondary, as the current version does not distinguish between the primary and secondary particles), the total dose, and the displacement damage energy deposition.  The tools allowed us to assess what kind of secondary particles are possible, their relative fluxes, and a comparable view of different shielding materials.  Such considerations will be important for satellite design and risk assessment for astronauts in long-term exploration and habitat.  However, this is just one example of many possible applications of MCNPX particle transport code. 

    ISNPS is currently evaluating several coupled monte carlo transport burnup packages for use in modeling reactor lifetime reactivity changes and actinide waste management concepts. MonteBurns 2.0 couples together MCNP5 with the burnup code Origen2.2. MCNPX 2.6c internally incorporates the CINDER90 burnup package within the code, simplifying its use. An in-house Simulink model used in conjunction with MCNP for modeling lifetime reactivity changes in space reactors is also being evaluated. The three packages combine work in similar fashions to model burnup, using the monte carlo transport code to determine the cross sections or reaction rates and feeding those rates to the burnup code for handling the production and depletion of nuclides in the burned materials. Monte carlo codes such as MCNP provide the capability to model virtually any system in almost any geometry granting great versatility to the packages. MonteBurns 2.0 uses MCNP5 to determine the one-group cross sections Origen2.2 requires as inputs to its models, while MCNPX 2.6c and the MCNP-Simulink model use one-group reaction rates calculated by the monte carlo code to input into the burnup package. MCNPX 2.6c can also produce 63 group fluxes for input to CINDER90 when the reaction rate cannot be calculated by the code.

    To determine which code is best suited for ISNPS’s computational modeling efforts the three code packages are currently being benchmarked against available experimental data from commercial pressurized water reactors. When the best suited transport-burnup package is determined it will then be used to calculate the operational lifetime for the SCoRe-S family of reactor concepts. The package will later be used to help model advanced fuel cycle concepts to support research in reducing the production of plutonium and other long-lived actinides in commercial power reactors and aid in the destruction of existing actinide wastes and plutonium stocks to ease future radioactive waste and proliferation challenges. Advanced mixed thorium oxide fuels, liquid metal cooled systems, and fast spectrum actinide burners are problematic to model with dedicated thermal light water reactor codes making a more general modeling package capable of modeling systems of almost any energy spectrum and material composition an invaluable tool for ISNPS’s research.

  • ORIGEN-ARP: Origen Automated Rapid Processing
    

    Origen-ARP is an isotopic depletion and decay analysis program designed to characterize nuclear fuel assemblies.  Part of the SCALE package, Origen-ARP provides for fast characterization of nuclear fuel under irradiation and decay conditions. ARP provides the Origen-S isotopic depletion code with problem dependent 3-group cross sections for its burnup calculations, allowing the package to model almost any commercial reactor currently in use.  Origen-ARP is being used at ISNPS to study the behavior of  MOX, (U,Th)O2, and inert matrix fuels for use next generation commercial nuclear reactors in reducing the production of  minor actinide, reducing the  mass and the volume of spent nuclear fuel.

  • SCALE 5: Standardized Computer Analyses for Licensing Evaluation version 5
    

    SCALE is a modular code that couples together functional modules, such as transport or depletion codes, allowing for combined analysis to be performed with a single code package. The program allows the user to use the same code package for handling neutronics, thermal, burnup, shielding, and other interrelated analysis in a coupled environment. SCALE 5 contains the 3D Monte Carlo transport codes KENOV and KENOVI, the burnup code Origen-S, TRITON, which links the 2D discrete ordinates code NWET with Origen-S, among many others. ISNPS uses SCALE for studies involving commercial power reactors and advanced nuclear Fuel Cycle and fuel management applications.

• CFD Analysis and Simulation
  
  
  • ANSYS
    

    ANSYS is a general purpose finite element modeling package for numerically solving a wide variety of mechanical problems. These problems include: static/dynamic structural analysis (both linear and non-linear), heat transfer and fluid problems, as well as acoustic and electromagnetic problems. In the past, obtaining all of the simulation capabilities needed for complex and demanding modeling scenarios frequently meant combining several different software packages. ANSYS Multiphysics provides the analysis industry's most comprehensive coupled physics tool combining structural, thermal, CFD, acoustic and electromagnetic simulation capabilities into a single software product. With ANSYS Multiphysics, you are getting the entire ANSYS simulation suite in one convenient package! ANSYS Multiphysics integrates the power of direct (matrix) and sequential (load vector) coupling to combine the appropriate "physical fields" required for accurate, reliable simulation results in applications ranging from cooling systems, power generation, to biotechnology and Micro Electro Mechanical Systems (MEMS). The software easily simulates complex thermal-mechanical, fluid-structural and electrostatic-structural interactions, and includes the complete range of powerful ANSYS iterative, direct and eigenvalue matrix solvers. ANSYS Mechanical includes a full complement of nonlinear and linear elements, material laws ranging from metal to rubber, and the most comprehensive set of solvers available. It can handle even the most complex assemblies especially those involving nonlinear contact and is the ideal choice for determining stresses, temperatures, displacements and contact pressure distributions on all your component and assembly designs.

    UNM-ISNPS developed and designed a novel, thermoelectric radioisotope power system for NASA's future space and planetary exploration missions (Figure 1). This system will not only have a specific electric power greater than 12 We/kg, which is more than twice that of the state-of-the-art GPHS Radioisotope Thermoelectric Generators (which power the Voyager, Galileo and Ulysses probes), but also an overall efficiency greater than 14%. Such high conversion efficiency would halve the amount of plutonium dioxide fuel needed for a given electric power requirement. The proposed advanced power system couples novel, segmented thermoelectric unicouples, based on advanced thermoelectric materials developed at JPL, to one or several standard General Purpose Heat Source bricks, and would be easily scalable to meet missions power requirement ranging from a few watts to hundreds of watts.

    During the course of this work, it was necessary to create and incorporate new numerical elements and new routines in the very powerful ANSYS 5.7 finite elements software to extend the capabilities of the software for ultimately modeling the entire radioisotope power system as well as the experimental setup. ANSYS has also been used at UNM-ISNPS to perform thermo-mechanical analyses of non-segmented (Figure 2) and segmented multicouples for RPSs.
    

    
    

    Figure 1. Skutterudite ThermoElectric Radioisotope Power System

    Figure 2.Temperature Contours and Displacements in a SiGe ThermoElectric Unicouple of a Power Conversion Assembly of a Jupiter Icy Moon Orbiter Spacecraft

    
    
  • MATLAB/SIMULINK
    

    ISNPS has developed a Dynamic simulation Model (DynMo-TE) of Space Reactor Power Systems (SRPSs) with Thermo-Electric (TE) conversion, and is currently developing the DynMo-CBC for simulating the dynamic behavior of SRPSs with Closed Brayton Cycle (CBC) engines using the SIMULINK® platform (http://www.mathworks.com/products/simulink). This platform offers an interactive graphical environment that allows rapid development of library blocks of the system components, thus components can be easily replaced or exchanged with little effort. Each component block has a number of input and output ports, and the blocks interact with each other by simply connecting these ports. Other input parameters necessary for the component blocks to operate are easily implemented through a customizable window or "mask," or by loading a MATLAB® input or "script" file. Another advantage of SIMULINK® is that is does not require the development and optimization of numerical schemes to deal with steep transients and the strong couplings inherent of the physical model equations. Through its integration with MATLAB® (http://www.mathworks.com/products/matlab), SIMULINK® has immediate access to an extensive range of tools for numerical computation, time integration, algorithm development, and data visualization and analysis. In addition to significantly reducing the development time and effort, such simulation capabilities make is easy to investigate the effect of replacing different types and/or using different designs of the system components such as pumps, heat pipes, etc. on the dynamic operation of the SRPS. SIMULINK® is also particularly well suited for studying the startup of the SRPS in orbit, and associated control scenario, and for developing appropriate adaptive control strategies for use on board the autonomous spacecraft in response to: (a) changes in system parameter over the long operation lifetime, caused by fuel burnup and degradation in materials properties foe example; and (b) unanticipated changes in environmental conditions, such as meteoroids impacts.

  • COSMOL/FEMLAB
    

    FEMLAB is a commercial source code based on the Finite Element Method for the computation of single or multi-physics applications. It has the ability to solve complicate systems more quickly and accurately using partial differential equations. This allows the program to simulate fluid flow and interaction in chemical and mechanical systems. ISNPS uses the newly released version, COMSOL MULTIPHYSICS-FEMLAB, which includes chemical engineering modules, to analyze single-phase fluid flow and heat transfer processes through micro-channels. With demands from the electronic, automotive and aerospace industries, smaller and more intricate mechanical actuators are required to obtain high performance in the restricted space. Micro-flow devices are also tending to be downscaled. The liquid flow and heat transfer characteristics in micro-channels are important in the design of such devices, but are still not well understood. To shed light on these effects, research is currently being conducted at ISNPS on the micro-effects, interaction between wall of micro-channel and fluid, using the FEMLAB finite element method code.

  • SOLIDWORKS / COSMOSWORKS
    

    SolidWorks® is a 3-dimensional design and mechanical engineering software used under license from Solidworks Corporation. It has been used extensively in the design and analysis of the Sectored, Compact Reactor (SCoRe) developed at UNM-ISNPS, a liquid-metal cooled space power reactor which prevents single-point failure within its cooling loops, and maintains subcriticality in the case of a launch abort incident. Originally modeled in SolidWorks®, this SCoRe concept model was used to assist and verify geometry inputs for the neutronics code, MCNP5 (add link to MCNP5). Since SolidWorks has the capability of entering physical properties for the structural materials in the SCoRe, such as density, the mass of the various components can be compared with the output from MCNP5 to verify the correct modeling of the system. SolidWorks® has also been helpful to ISNPS engineers in designing nuclear radiation shields and other space nuclear reactors (Figure 3), such as heat pipe-cooled reactors for which the main challenge is the routing of the relatively large number of reactor heat pipes around or through the radiation shield (Figure 4), and their coupling to the energy conversion units in the space power system. The mechanical animation capability of SolidWorks® has also been used successfully to obtain snapshots of space reactor power systems with segmented and deployable radiation panels, at different times during the deployment process in space orbit.

    COSMOSWorks is a powerful, easy-to-use design validation and optimization software fully embedded within the SolidWorks® software. COSMOSWorks uses the Finite Element Method to provide one-screen solutions for stress, frequency, buckling, thermal and optimization studies. One component of this package, FloWorks, has been used to verify fluid dynamics and heat transfer calculations within both the S^4 and SCoRe nuclear reactors. In the S^4 reactor, FloWorks was used to optimize coolant channels placement and inlet and outlet plenum designs for minimum pressure loss and maximum heat transfer. Pertaining to the SCoRe reactors, FloWorks was used to verify pressure drops within the coolant inlet annulus.

    Figure 3.Isometric View of the Scalable AMTEC Integrated Reactor Space Power System (SAIRS-C).

    Figure 4. Cross-sectional and Isometric Views of the HP-STMC Reactor and Radiation Shield

    
    
• System Simulation & Transient Analysis
  
  
  • HPTAM
    

    The Heat Pipe Transient Analysis Model "HPTAM" (Figure 12) was developed at UNM-ISNPS for simulating the transient and steady-state operation, as well as the startup from a frozen state of alkali-metal and low-temperature heat pipes. Heat pipes are used in many space nuclear reactor power systems for transporting the reactor's thermal power to the converters (lithium or sodium heat pipes), transporting the waste heat from the cold side of the converters to the heat rejection radiator (sodium or potassium heat pipes), and/or for spreading the heat in finned radiator surfaces (potassium, rubidium or water radiator heat pipes). HPTAM has been the subject of continuous upgrade and verification since the early nineties using experimental data for water, sodium (Figure 13), and lithium (Figure 14) heat pipes. The fully 2-dimensional and transient HPTAM model (Figure 12) uses the Darcy's extended flow equations to model the liquid flow in the porous wick and annular space, in conjunction with a fixed-grid, homogeneous enthalpy method to calculate the change of phase in these zones. HPTAM calculates the radius of curvature of the liquid meniscus in the surface pores of the wick, and couples the vapor and wick regions with appropriate momentum and enthalpy jump conditions, which allows HPTAM to predict the capillary limit, partial recession of liquid in the evaporator wick, and pooling of excess liquid in the vapor core. HPTAM also calculates the rates of sublimation of the frozen working fluid in the evaporator and resolidification in the adiabatic and condenser sections early during the startup from a frozen state, and simulates the free-molecule, transition, and continuum vapor flow regimes using the Dusty Gas Model.

    Figure 12. Block-Diagram of HPTAM's Physical Model

    Figure 13. Axial Distributions of Wall and Vapor Temperatures during the Startup of a Sodium Heat Pipe (Faghri et al., 1991)

    Figure 14. Wall Temperature during Startup of Li Heat Pipe; Symbols Represent TC data at Particular Times (Reid, Sena, Merrigan, Elder and Martinez 1999)

    
    
  • AMTEC Performance and Evaluation Analysis Model (APEAM)
    

    In the 1990's, UNM-ISNPS participated in a comprehensive testing and modeling program with the Air Force Research Laboratory's (AFRL) Space Vehicles Directorate. The objective of this program was to advance the technology of vapor anode, multi-tube Alkali-Metal Thermal-to-Electric Converters (AMTECs) for flight on future space

    Figure 9. Cross-section views of a typical vapor anode AMTEC converter with 6 BASE tubes connected in series (not to scale)

    Figure 10. Predicted heat flow and structure temperatures in the PX-3A converter when operating at a peak electrical power of 4.7 We (experimental measurements are shown in parentheses

    Figure 11. I-V Characteristic and Electric Power Output of PX-3G Cell #1 in Ground-Demo

    missions. Each vapor anode, multi-tube PX-series AMTEC cell uses between 5 and 8 Beta''-Alumina Solid Electrolyte (BASE) tubes, connected electrically in series (Figure 9). The TiN or WRh anode and cathode porous electrodes are covered with molybdenum mesh current collectors, to minimize internal electric losses. A two-dimensional, integrated AMTEC Performance and Evaluation Analysis Model (APEAM) was developed at ISNPS, to support ongoing tests at AFRL and improve the design and performance of sodium and potassium PX-type AMTEC cells. This integrated cell model consists of four major components:
    

    • (a) an alkali-metal vapor pressure loss model, which calculates the low vapor pressure at the interfaces between the cathode electrode and the BASE tube;
      
    • (b) a two-dimensional radiation/ conduction heat transfer model, which accounts for all heat exchanges between the different components of the cell and calculates the temperatures throughout the cell (Figure 10);
      
    • (c) a cell electrochemical model, which calculates the effective potential developed across the BASE, due to the isothermal expansion of alkali metal ions; and
      
    • (d) a two-dimensional electric circuit model, which determines the electrical resistances of the BASE, electrodes, current collectors, and conductor leads to the external load, and calculates the cell's electrical potentials, electrode current density, and the cell's total electrical current. APEAM has been successfully benchmarked against experimental data of individual PX-type converters and of an eight-cell power generator ground demonstration that were electrically heated (Figure 11).

    
    
  • DynMo - TE
    

    ISNPS has developed, using the SIMULINK® platform, the Dynamic simulation Model (DynMo-TE) of Space Reactor Power Systems (SRPSs) that use a Sectored, Compact Reactor (SCoRe) designed for avoidance of a single point failure, and Thermo-Electric (TE) conversion (Figure 5). DynMo-TE (Figure 6) is comprised of a number of coupled physical models:
    

    1. The SCoRe Model couples a six-points kinetics model to a thermal-hydraulic model of the reactor core (Figure 6). The former calculates the reactor fission power, subject to the external reactivity insertion at a user specified rate and the temperature-reactivity feedback for the liquid lithium and UN fuel, and the Doppler reactivity feedback, when applicable.
    2. The Secondary and Primary Loop Models are coupled thermally in a PCA and a pumps TCA (Figure 6). The coolant flow rates and pressure losses in the primary and secondary loops are calculated using the mass and momentum balance equations, at the points where the pressure loss demand-curves intersect the calculated pressure head supply-curves of the EM pumps.
    3. The Electro-Magnetic (EM) Pump Model calculates the pressure head supply-curve as functions of time and temperature, as function of the thermal and electrical resistivities of the duct wall material, coolant and copper buses, and the DC voltage and current supplied by the pumps TCA. The EM pumps use permanent magnets that are thermally insulated from the coolant ducts and maintained well below their Curie point. The magnet remains saturated by the secondary magnetic flux generated by the electric current passing through the pump ducts.
    4. The Radiator Panel Thermal-Hydraulic Model couples the 3 radiator segments (one forward and 2 in the rear) hydraulically in parallel (Figure 5), and discretizes the inlet and outlet flow channels of each segment into small axial sections. Each section comprises a small number (5 or 6) of rubidium heat pipes with C-C armor and C-C fins, and solves the coupled momentum and energy balance equations for the coolant flow rate and temperature. The model also calculates the temperature drops in the channel walls and the structure of the evaporator section of the heat pipes. The radiator model is coupled to a heat pipe model, to calculate the vapor flow and temperature drops in the heat pipes walls and C-C fins. The heat pipe model also calculates the sonic, capillary, entrainment, and incipient boiling limits.
    5. The Power Conversion Assembly (PCA) and Pumps TE Converter Assembly (TCA) Models both use a transient performance and optimization model of the SiGe unicouples. This model is also capable of predicting the performance of segmented TEs, with up to 3 different materials in each leg.
    6. Accumulator Model: To accommodate the volume changes in the liquid metal coolant in the secondary and primary loops during transient operation, each loop is equipped with a bellow type accumulator (Figure 6). The accumulator model accounts for the stiffness of the bellows, in addition to that of the compression spring. For operation redundancy, the cavity above the bellows is filled with inert gas to support the spring and the bellows in adjusting the coolant pressure in the loops. The accumulator model calculates the transient changes in the coolant volume in the accumulator, the compression length of the helical spring, and the coolant pressure in the loop.

    
    DynMo-TE has been used successfully to optimize the design of the SRPS, investigate the effect of using different combinations of alkali metal coolants (Li, Na, and NaK) in the primary and secondary loops, study the startup transient and propose a startup scenario to minimize the amount of startup batteries needed, and investigate the load-following operation of the SRPS (Figure 7). DynMo-TE is particularly well suited for developing safe startup procedures and schemes for adaptive and autonomous operation and control of the SRPS.

    
    

    Figure 5. A Layout of SCoRe-TE Space Nuclear Reactor Power System

    Figure 6. DynMo-TE for Dynamic Simulation of SCoRe-TE SRPS

    Figure 7. Steady-state Operation Map of SCoRe-TE Power System

    
    
  • DynMo - CBC
    

    ISNPS is currently developing, using the SIMULINK® platform, the Dynamic simulation Model (DynMo-CBC) of Space Reactor Power Systems (SRPSs) that use a gas-cooled reactor designed for avoidance of a single point failure, and Closed Brayton Cycle (CBC) engines (Figure 8). This model shares many of the same characteristics as DynMo-TE. Figure 1 presents the layout and the nominal full-power performance parameters of a typical SRPS with 3 CBC loops. The sectored, fission reactor, cooled with He-Xe, a Pellet Bed Reactor (PeBR) or Submersion-Subcritical Safe Space (S^4) reactor, is divided into three identical sectors loaded with nuclear fuel elements. These sectors are neutronically and thermally coupled, but hydraulically decoupled. Each reactor sector provides thermal power to a CBC loop with a single-shaft radial-flow Compressor-Generator-Turbine unit, a recuperator and a gas cooler. The gas coolers in the three CBC loops heat up the molten NaK-78 in the secondary loops, which transports the waste heat to water heat pipes radiator panels. The panels, two per CBC loop, are hydraulically coupled in parallel to reduce pressure losses. The water heat pipes have carbon-carbon (C-C) fins to extend the surface area for heat rejection, and C-C armor to protect against the impact of meteoroids. The NaK-78 in the secondary loops of the radiator panels is circulated using Alternative Linear Induction Pumps (ALIPs). The operating parameters of the CBC engines presented in Figure 8 account for the mechanical losses in the bearings, the electrical losses in the alternator windings, and the electromagnetic electrical losses in the electrical generator.

    

    The system layout in Figure 8, with a sectored nuclear reactor core, avoids single point failures both in reactor cooling and energy conversion. With a failure of one of the CBC engines, a loss-of-cooling, or a break in one of the primary gas loops, the present space nuclear power system continues to operate with two CBC engines, but at reduced electrical and reactor thermal powers. The fission power generated in the reactor sector of the failed CBC loop is transported by conduction and / or by thermal radiation to the dividers between the sectors and removed by forced convection to the circulating gas in the two adjacent sectors. DynMo-CBC will be useful for optimizing the design of the SRPS, studying the startup transient, investigating the load-following operation of the SRPS, and developing safe startup procedures and schemes for adaptive and autonomous operation and control of the SRPS.

    Figure 8. DynMo-CBC for Dynamic Simulation of Multiple-Loop Brayton SRPS

    
    
  • UNM Heat Pipe Model (UNM-HP)

    UNM-HP is a steady-state performance and optimization model of fully-thawed alkali liquid metals (K, Na, and Li) and water heat pipes. It calculates the vapor pressure losses in the various sections of the heat pipe (evaporator, adiabatic, and condenser), when vapor flow is dominated by friction rather than inertia forces (vapor Mach number < 0.5). The calculated vapor pressure loss determines the temperature drop in the vapor core along the heat pipe. This temperature drop is added to the conduction temperature drops in the wall, liquid annulus, and the liquid saturated wick, both in the condenser and evaporator sections. The sum of these temperature drops gives the total temperature drop along the heat pipe. UNM-HP accounts for both the inertia and friction forces in the vapor flow, and the vapor remains saturated. This heat pipe model has been verified successfully using experimental data of the vapor temperature along a short sodium heat pipe (0.7 m) operated at different powers and Mach numbers up to 0.4 (Ivanovskii et al., 1982; see Figures 15a and 15b).

    Figure 15. Comparison of UNM-HP Predictions with Measured Vapor Temperatures in a Sodium Heat Pipe at Mach numbers of 0.15 and 0.30 (Ivanovskii et al. 1982)

    Figure 16. Operation Limits of Potassium Heat Pipe in SAIRS-C Radiator

    The UNM-HP model has been used successfully in the design and thermal analysis of the heat pipes-cooled nuclear reactors and heat rejection radiator panels of the HP-STMC and SAIRS power systems developed at UNM-ISNPS. In such designs, the local radial and axial vapor mass fluxes of working fluid in the heat pipe are calculated from the local energy balance, which accounts for the heat input to the evaporator section and the heat rejection along the condenser section. UNM-HP also accounts for the heat conduction in the C-C fins of the radiator heat pipes and calculates various heat pipes operation limits; namely, the viscous, sonic, capillary, entrainment, and incipient boiling, to determine the useful operation domain bound by these limits. Figures 16a and 16b compare the predictions of the UNM-HP model of the operation limits of the radiator heat pipes in SAIRS-C with those obtained using the widely used HTPIPE model, developed by Los Alamos National Laboratory (Woloshun et al., 1989). As these figures indicate, the predictions of the sonic and the capillary/wicking limits by the two models are in good agreement. Figure 17 shows the temperature contours of the SAIRS radiator surfaces predicted by the model at nominal operation. The Scalable AMTEC Integrated Reactor Space Power Systems (SAIRS, Figure 3) developed at UNM-ISNPS all use D-shaped potassium heat pipes radiators with C-C armor and fins.

    Figure 17. Surface Temperature Contours of SAIRS Radiator C-C Armor/Fin

    
    

Recent Publications

• 2008
• 2007
• 2006
• 2005
• 2004
• 2003
• 2002

Back to Top

Last updated 4/6/07. ©2006-2007 ISNPS-UNM.